The search for alternate forms of energy has in recent years drawn a great deal of attention to technology associated with the recovery of uranium. In order to produce uranium in a usable form, the uranium-containing minerals must be treated in order to extract and concentrate the uranium values. Depending on considerations such as the quality of ore, the geology and hydrology of the mineral formation, location and distribution of the ore, and environmental impact, the ore may be mined and milled to expose the valuable uranium for contact with a leaching solution, or, if the ore body is not so rich so as to justify expensive mining and scarring of the earth's surface, the uranium may be leached from the ore in situ.
Both methods of uranium recovery require a leaching process by which uranium values are extracted from the gangue, i.e. worthless rock or vein matter in which valuable metals or minerals occur, by means of a leaching fluid or lixiviant. In the case of in situ leaching the leaching fluid is introduced to the ore body through a predetermined pattern of injection wells. The leaching solution, which may be acidic or alkaline depending on the nature of the ore, consists essentially of an oxidant, e.g. O.sub.2, H.sub.2 O.sub.2 or NaCl.sub.3, and a leaching agent such as Na.sub.2 CO.sub.3 /NaHCO.sub.3 or Na.sub.2 SO.sub.4 /NaHSO.sub.4. Carbonate leaching is an especially attractive method for several reasons including selectivity of the reaction, purity of the resultant leach liquor, handling ease, etc.
Upon contact with the uranium ore insoluble uranite (UO.sub.2) is oxidized by the oxidant in the lixiviant from a tetravalent to a hexavalent leachable state (UO.sub.3) according to the following equation, EQU UO.sub.2 +H.sub.2 O.sub.2 =UO.sub.3 +H.sub.2 O (1)
which in turn, reacts as in equation (2) below, with the leaching agent to form a soluble uranium complex; EQU UO.sub.3 +Na.sub.2 CO.sub.3 +2NaHCO.sub.3 =Na.sub.4 UO.sub.2 (CO.sub.3).sub.3 +H.sub.2 O (2)
The resulting uranium-enriched solution (pregnant leachate) is then retrieved through a pattern of production wells for separation of the uranium ion complex by means of ion exchange processes.
Ion exchange is, generally speaking, a reversible exchange of ions between a liquid phase and a solid phase which is not accompanied by any substantial change in the basic solid ion-exchange resin structure. Ion-exchange resins consist of three-dimensional covalent networks based on carbon-carbon bonding, to which are attached bound ions. The covalent cross-linkages preserve structural integrity of the materials in service, while the bound ions constitute ion-exchange sites thereby causing the ion-exchange resin to act as an insoluble acid or base which forms an insoluble product with the reactant ion. There are two general types of ion exchange resins --cationic and anionic--depending on the type of ion that is exchanged between the liquid and solid. Thus, an ion exchange resin which removes and releases negatively charged ions, such as uranyl-trisulfate and uranyl-tricarbonate ions, is considered an anion exchange resin. Ion exchange resins are semi-rigid gels generally prepared as spherical beads.
When a solution of uranium-containing anions is passed over an anion exchange resin the reaction proceeds according to the following equations in the case of uranyl-trisulfate and uranyl-tricarbonate, respectively: EQU 4R.sup.+ X.sup.- +UO.sub.2 (CO.sub.3).sub.3.sup.-4 .revreaction.R.sub.4 [UO.sub.2 (CO.sub.3).sub.3 ]+4X.sup.- ( 3) EQU 4R.sup.+ X.sup.-+UO.sub.2 (SO.sub.4).sub.3.sup.-4 .revreaction.R.sub.4 [UO.sub.2 (SO.sub.4).sub.3 ]+4X.sup.- ( 4)
wherein R.sup.+ designates the active ion exchange sites on the resin and X.sup.- may be any nucleophilic ion, such as a halide. Based on the law of mass action, the above reaction can be made to proceed in the desired direction by merely increasing the concentration of the ion shown as the free radical.
Therefore, in order to capture the uranium-containing ion on the resin, the concentration of the uranium ions is increased by introducing the leachate to the resin thereby causing the reactions to proceed in the direction of the top arrows. Once the leachate has been passed over the resin, the uranyl complexes are retained on the resin while the rest of the leachate, along with the impurities mixed therewith as a consequence of the leaching operation, may be recovered and treated for recycle without fear of losing the precious uranium.
The uranium ions can then be recovered from the resin by passing a solution containing, for example, chloride ions through the resin so that the reactions are displaced in the direction of the lower arrows. This solution, called an eluant, may contain NaCl and HCl in order to ensure that the appropriate exchange takes place.
The typical anion exchange resin is described as a quarternary ammonium ion-exchange resin. A type I strong base anion exchange resin, such as Dowex MSA-1 and Dowex 2K, is particularly useful because of its effectiveness in capturing the uranyl complex ions as they are contacted with the resin as part of the leachate.
While there are two methods which can be used to contact the uranium-containing solution with the anion-exchange resin, i.e. the batch method and the column method, for commercial purposes the "column method" is preferred. The column method includes passing the leachate through an elongated column which contains the resin, followed by the passage of a second solution through the column which elutes the uranium-containing complex ions from the resin. The latter step is usually referred to as the elution stage of the cycle. Elution solutions necessarily include ions which, as applied to the discussion above, would replace the uranyl complex on the resin site and cause the uranyl complex to go into solution. A typical elution solution may include concentrations of NaCl or (NH.sub.4).sub.2 CO.sub.3 thereby providing Cl.sup.- or CO.sub.3.sup.-2, respectively, as the replacement ions.
The uranium pregnant eluate is then collected in vessels for precipitation preferably by reacting the uranyl complex with hydrogen peroxide in an acid solution to form a hydrated uranium peroxide product, e.g. UO.sub.4.XH.sub.2 O.
The resulting precipitate, yellow-cake slurry, is pumped to storage tanks for settling and decanting. After the slurry is settled and the barren solution decanted the slurry is pumped to a vacuum dryer where it is dried to yellow-cake powder.
Although uranium recovery processes such as described above are currently being used, practical problems still exist, including in the use of ion-exchange resins. Specifically, the efficiency of uranium ion capture by the anion exchange resin decreases with use because of the continual buildup of contaminants on the charged ionic resin sites which cannot be removed during normal elution and/or washing of the resin. In the case of both mining and in situ leaching processes wherein the gangue materials are rich in sulfides, such as in pyrites, the ion-exchange resin contaminahts include polythionates, S.sub.n O.sub.6 (wherein n.gtoreq.4). These partially-oxidized sulfur compounds are strong competitors for the exchange sites of the resins and, moreover, are not easily removed from the resin during the elution step. Continual buildup of the undesirable polythionates on the ion exchange resin reduces the number of sites available for the capture of uranyl complexes to such an extent that the expensive exchange resin must be replaced because of excessive uranium "leakage".
In the past, it has been known in the art of anion exchange resins in general to regenerate such resins by contacting them directly with aqueous solutions of an alkali, such as sodium hydroxide, ammonium hydroxide, and sodium carbonate (see U.S. Pat. No. 2,669,713) or with a liquid slurry of solid particles of hydroxide of a metal from the group consisting of calcium, magnesium, barium and strontium (see U.S. Pat. No. 2,507,992). In the case of weak base anion exchange resins, it is known that regeneration may be achieved by contacting the resin with an aqueous lime slurry (see U.S. Pat. No. 3,429,835).
Defouling anion exchange resins of iron and iron oxide is disclosed in U.S. Pat. Nos. 3,078,228 and 3,267,883, respectively. As for iron contaminants, the resin is backwashed with a solution containing hydrosulfites, whereas the iron oxide may be removed by contacting the resin with a chelating agent taken from the class consisting of sodium and potassium polyphosphates and ethylenediamine tetraacetic acid.
Rejuvenation of strong base anion exchange resins contaminated with molybdenum, titanium, zirconium, bismuth, silicon (as silicate), phosphorus (as phosphate), hafnium, tungsten, and, to a lesser extent, iron, chromium, lead, tin, and manganese, is described in U.S. Pat. No. 3,252,920. The cleansing process includes the steps of treating the resin, preferably, as follows:
(1) Acidified nitrate solution PA0 (2) Wash (water) PA0 (3) Alkaline carbonate solution PA0 (4) Moderately strong alkaline solution PA0 (5) Sodium chloride solution PA0 (6) Dilute or moderately strong mineral acid solution (other than phosphoric acid) PA0 (7) Rinse (water)
In U.S. Pat. No. 2,812,312 a method for regenerating ion exchange resins used to remove higher alkyl sulfates from an aqueous solution by use of a strong mineral acid such as sulfuric acid H.sub.2 SO.sub.4 is disclosed. Other sulfur compounds, including sulfites, have in the past been contemplated as potential reagents for resin regeneration. See E. A. Nugent, "The Chemistry of the Poisons Associated With the Ion-exchange Process," The South African Industrial Chemist, Vol. 10, pg. 282-290, November, 1956, wherein sulfites are included among a list of materials in the treatment of resin poisons of the polythionite type. Caustic soda is said to be the compound generally employed for the purpose. In any event, it has never been suggested to use sulfite in a uranium recovery process, nor has there been a recommended dosage which in theory and in practice effectively removes essentially all the polythionates captured on the resin.
Of interest also to the invention disclosed herein, is a paper presented by Velu Annamalai and Frances X. McGarvey at the 55th Annual Fall Technical Conference and Exhibition of the Society of Petroleum Engineers of AIME, held in Dallas, Tex., Sept. 21-24, 1980 (Published article "Operating Experience in the Recovery of Uranium at the Pawnee and Zamzaw Sites," SPE 9507, 1980). The authors state that even though a review of polythionic chemistry indicates that combinations of oxidizing agents and acids might be expected to return the resin to a practical exchange capacity, efforts to date to eliminate polythionates based on such combinations have not been successful. Specifically, a chemical treatment of the resin using 5% caustic, a combination of caustic and salt, and lowering the treatment solution to low pH with diluted hydrochloric acid resulted in some initial improvement in the resin loading capability, but did not produce any substantial improvement in the overall performance of the resin. Further efforts included the use of hydrogen peroxide in attempting to keep the resin clear, but resulted in a physical degradation of the resin due to a general attack on the hydrocarbon chain by the H.sub.2 O.sub.2. Consequently, the resin exhibited an increase in water retention, swelling, reduction in volume capacity, a general reduction in mechanical strength, and a formation of N-oxide compounds which causes de-quaternization of the strong base groups and a concomitant reduction in uranium holding capacity. The culmination of this work suggests that, even though the study of polythionic chemistry might lead the novice to attempt to rid the exchange resins of polythionate poisoning by use of strong nucleophilic ion species, experience and testing have contraindicated this approach.
Attempts to eliminate polythionates from ion exchange resins have also led, with some success, to treatment of the resin with cyanide, but the use of cyanide creates obvious safety and environmental problems.
It is a primary concern of the present invention to provide a process by which an anion exchange resin used in the recovery of uranium may be kept continuously free of polythionate poisoning.
A further object of this invention is to provide a process wherein the resin does not suffer physical degradation as a result of polythionate cleansing and which presents no safety and/or environmental problems.